US-APWR

US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 6
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
LTD.
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UAP-HF-08010
Presenter
Andrew B. Johnson
Principal Engineer
Mitsubishi Nuclear Energy Systems, Inc.
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Contents
1. Overview of Chapter
,/ Title of Chapter
/ Scope of Chapter
2. Design Features
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UAP-HF-08010-2
Overview of Chapter
>Title of Chapter
Chapter 6:
ENGINEERED SAFETY FEATURES (ESFs)
;Scope of Chapter
This chapter includes the following ESFs:
* 6.1 Engineered Safety Features Material
* 6.2 Containment Systems
0 6.3 Emergency Core Cooling Systems (ECCS)
• 6.4 Habitability Systems
• 6.5 Fission Product Removal and Control Systems
a 6.6 Inservice Inspection of Class 2 and 3 Components
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2. Design Features
> 6.2 Containment Systems
Containment systems consists of followings:
* Containment Structure (PCCV)
* Containment Spray System
* Containment Isolation System
- Containment hydrogen monitoring and
control system
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2. Design Features
> Containment Systems (Cont'd)
/ Containment Function
The containment is designed as an essentially
leak-tight barrier that will safely and reliably
accommodate calculated temperature and
pressure conditions resulting from loss-ofcoolant accident, or main steam line break.
Major Design Parameter
Type
Design Pressure
PrestressedConcrete Containment Vessel
(PCCV)with Carbon Steel Liner
68 psig
Design Temperature
300 deg. F
Design Leakage Rate
0. 1% airmass Iday
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2. Design Features
Containment Systems (Cont'd)
" Containment Function (Cont'd)
Example analysis result (largebreak LOCA)
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2. Design Features
Containment Systems (Cont 'd)
V
Containment Heat Removal
•4 Independent trains
-Automatic initiationby
ContainmentSpray Signal
-Pumps and heat
exchangers used for RHR
functions during shutdown
-Common Spray Ring
Header
INSIDETHE CONTAINMENT I OUTSIDETHE CONTAINMENT
Spray Ring Header
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Take suction from in-containment RWSP
Note: Red portions are common part for CSS and RHRS
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2. Design Features
6.3 Emergency Core Cooling Systems (ECCS)
.-
4 Independent trains
-- ----
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Automatic initiationby
Safety Injection Signal
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for Safe Shutdown
Advanced
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2. Design Features
Emergency Core Cooling Systems (EC( CS) (Cont'd)
v' Advanced Accumulator
*
*
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Automatic switching of injection flow rate by flow damper
Integrates function of low head injection sy stem
Long accumulator injection time allows Ionger time for safety
injection pump to start
Advanced Accumulator
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2. Design Features
;6.4 Habitability Systems
The habitability systems allow operators to remain safely inside
the control room envelope (CRE), that includes the main control
room (MCR), and take the actions necessary to manage and
control the plant under abnormal plant conditions, including a
LOCA.
VMCR HVAC System
a 2 x 100% MCR emergency filtration units
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4 x 50% MCR air handling units
Air tight isolation dampers
Two emergency modes
Pressurization mode ; during an accident with
radiological releases.
Isolation mode ; during a toxic gas event
Automatic initiation by the MCR isolation signal
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2. Design Features
Habitability System (Cont'd)
trol Room Envelope
Air Tight
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UAP-HF-08010-12
2. Design Features
6.5 Fission Product Removal and Control Systems
,,The fission product removal systems remove fission products that
are released from the reactor core as a result of postulated
accidents.
,/The containment controls the leakage of fission products to ensure
that the leakage rate from the containment is below limits.
,/The US-APWR fission product removal and control systems are
as follows:
"Containment spray system
" Containment
"Annulus emergency exhaust system
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UAP-HF-08010-13
2. Design Features
!NAPSW
Fission Product Removal and Control System (Cont'd)
Fission product removal effects differ with the chemical forms of the radioactive iodine. The assumed
chemical forms are noble gas, elemental iodine, organic iodine, and particulate (aerosol). The fission
product removal effects in the US-APWR containment under accident conditions are the following:
Mechanism
Containment Spray
Noble Gas
Not Applicable
Elemental Iodine
Organic Iodine
Particulate (Aerosol)
Slight effect, No credit
applied (Not t)
Not Applicable
Not Applicable
Applicable
(Powers natural
deposition model
(NUREG/CR-6189):
10t percentile)
Applicable
(Based on SRP 6.5.2)
Natural Deposition
Not Applicable
Applicable (Note 2)
(Based on SRP 6.5.2)
Radioactive Decay
Applicable
Applicable
Applicable
Applicable
Containment Leakage
Applicable
(Based on Technical
Specifications)
Applicable
(Based on Technical
Specifications)
Applicable
(Based on Technical
Specifications)
Applicable
(Based on Technical
Specifications)
Not Applicable
Not Applicable
(Note3)
Annulus Emergency
Exhaust System
Not Applicable
III_
Applicable
I_(HEPA filter)
Notes:
1.
2.
3.
13
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The CSS with NaTB baskets is expected to achieve a pH of at least 7 in the RWSP. Thus, the CSS can remove elemental iodine slightly.
Therefore, we assume that the CSS does not remove elemental iodine.
The CSS removal effects contain the removal effect by naturaldeposition. Because the removal effects for elemental iodine by the CSS is
not credited, the removal effects for elemental iodine by naturaldepositioncan be creditedin not only the sprayed region, but also the
unsprayed region.
Containment Leakage to the penetration areas is treatedby the annuls emergency exhaust system
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UAP-HF-08010-14
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 7, 8 and 18
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
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WEVYN4INDUSIR1ES, LTD.
UAP-HF-08011
Contents
> Chapter 7:
Instrumentation and Controls (I&C)
> Chapter 8:
Electric Power
> Chapter 18: Human Factor Engineering (HFE)
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For each Chapter
1. Content Overview
2. System Descriptions
3. Analysis and Evaluations
> Summary
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Chapter 7- Content Overview
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> Chapter 7 includes the following descriptions:
,/ All safety related I&C systems
,/ Non-safety I&C systems which are important in
maintaining safe normal operating conditions and which
support abnormal plant conditions
,/ Intra and inter system data, communications
> Descriptions focus on features related to:
/ Performance
/ Reliability
/ Maintainability
/ Failure modes
> Format based on RG 1.206
> Content based on RG 1.206 and SRP
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I&C System Overview (7.1)
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Common digital microprocessor based platform for
safety and non-safety I&C (no electro-mechanical relays)
Diverse Actuation System based on analog technology
Complete four train redundancy for safety i&C with each
division in separate fire area
Distributed architecture for non-safety I&C with
redundancy
Fully multiplexed and duplicated signal transmission
networks from local areas to I&C equipment rooms and
between I&C systems/components
Fully computerized Main Control Room and Remote
Shutdown Room with no reliance on local controls
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UAP-HF-08011-4
i&C System Overview (7.1)
The digital I&C and HSI systems for the US-APWR are
essentially the same as the I&C and HSI systems for
nuclear plants in Japan
V First installation for non-safety digital I&C in 1987
/ Average 10 years operation for five operating plants
V Applied to all non-safety I&C, 50 applications per plant
/ Over 20 million hours.total operating experience
v/ No un-expected shut down caused by I&C since 1992
/ No system malfunction caused by S/W or H/W failure
/ The same digital platform is currently being applied to safety
and HSI systems of the Japanese APWR, and safety and HSI
systems currently being implemented for plant modernization.
• First safety and HSI application: Tomari #3, C/O 2009
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UAP-HF-08011-5
UAP-HF-0801 1-5
Main Control Room (7.1)
Operator
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Safety VDU1
Alarm VDU
Operation VDU (Non-Safety)
UAP-HF-08011-6
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Chapter 7
-
Analysis & Evaluations
> Chapter 7 descriptions include following design details:
,/ Redundancy, separation and isolation
V Data communication independence and performance
V Maintenance and operating bypasses
V Test coverage (self-test and manual)
V Access controls and cyber security
/ Failure modes and effects
V Coping with common cause failures
/ Hardware and software quality (Software life cycle)
,/ Hardware qualification and reliability
> Chapter 7 describes the following design processes:
v/ Setpoint determination
$ Software life cycle (basic and application)
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UAP-HF-08011-12
MELTAC Platform
Mitsubishi Electric Total Advanced Controller
Simple Design
V Modular and Structured Architecture
/ Single Task execution
V Cyclical Processing with No Interrupts
> Quality Assurance and Control
V Designed specifically for Nuclear Applications
v/ Under'control of Nuclear QA/QC
V Fully owned and life cycle managed by Mitsubishi
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UAP-HF-08011-13
Chapter 8 - Content Overview
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Chapter 8 includes the following descriptions:
V On-site safety related AC and DC power systems
V On-site non-safety AC and DC power systems
v/ Interface to off-site power distribution system
) Descriptions focus on features related to:
V Reliability
/
Maintainability
/ Failure modes
Format based on RG 1.206
Content based on RG 1.206 and SRP
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Power System Overview (8.1)
Transmission Systeml
UAT1,2: Unit Auxiliary Transformer 65MVA
UAT3,4 : Unit Auxiliary Transformer 53MVA
RATI 2 Reserve Auxiliary Transformer 65MVA
RAT3,4 : Reserve Auxiliary Transformer 53MVA
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Non Class I E (Non Safety
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UAP-HF-08011-15
Offsite Power System (8.2)
> Design Features
,The two (2) sources of offsite power provided.
a) Main Transformer through Unit Auxiliary
Transformers (UAT)
b) Reserve Auxiliary Transformer (RAT)
-/The two (2) offsite power supply circuits are
independent and physically separated.
,/Both offsite power supply circuits have enough
capacity to achieve their safety related function
during a Design Basis Event (DBE) and meet the
requirement of the applicable GDC's.
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UAP-HF-08011-16
Onsite AC Power System (8.3)
Design Features
1 E AC electrical power system consists of four
(4) separate trains. Each train includes one Class 1 E
Emergency Power Source (EPS)
,/On-Line Maintenance of any EPS is allowed with
Single-Failure Criterion remaining satisfied
V"Permanent" buses supplied from Alternate AC
Power Source (AAC) are provided
V/Non-safety related loads are not supplied from class
1E buses. Required non-safety related loads are
supplied from AAC in LOOP condition
,/AACs provide power to all electrical loads that are
required to bring and maintain the unit in safeshutdown mode upon the SBO
V/Class
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UAP-HF-08011-17
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UAP-HF-OROI
UAP-HF-08011-17
Gas Turbine Generator (8.3)
Gas turbine
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Power seotlon
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Gear box
Coupling
MHI selected
Gas Turbine Generators
for EPS and ACC
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Why Gas Turbine Generators
,/GT/G has been selected based on reliability and maintainability
improvements when compared to DG
Gas Turbine Generator
Diesel Generator
Compact
Large
Not Required
Required
1/3 the parts of a DG
Complex
Large Scale OverhaullieRqrd
Once or twice during plant
life
Periodic Overhaul
Required
Reliability (failureldemand)
104 based on Japanese
experience
10-3
40 sec
10 sec
Space
Cooling Water
Routine Maintainability
Starting Time
V'Longer start time of GT/G is accommodated by the Advanced
KuTmm!
Accumulator design of US-APWR which allows 100 sec
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UAP -HF-08011-19
Station Blackout (8.4)
Basic Concept for Coping with SBO
AACs are available in the event of SBO, when all offsite
power sources and EPSs are not available to bring the unit
to a safe shutdown condition and maintain that status
/The
Design Basis
,/AACs of a different type (Starting System, Capacity etc.)
and are provided to minimize the potential for common
mode failure with either the offsite power or the EPS system
v/ The AAC is a non-class 1E gas turbine-generator package
connected to a 6.9kV AC "Permanent" bus
, The AAC supplies power to loads on any class 1 E bus
through tie line circuits during SBO
V The AAC supplies power to loads for 8 hours during SBO
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UAP-HF-08011-20
Chapter 8 - Analysis & Evaluations
kAPW
Chapter 8 descriptions include the following
design details:
,/ Redundancy, separation and isolation
v/ Failure modes and effects
V Hardware quality
, Hardware qualification and reliability
>Chapter 8 describes the following design
processes:
Class I E qualification and tests of Gas Turbine
Generator
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UEAP-HF-080f1 1-21
Chapter 18 - Content Overview
Chapter 18 includes the following descriptions:
,/Human Factors Engineering process
v/ Human Systems Interface design features
Descriptions focus on features and processes
intended to:
/ Enhance human performance
/ Reduce potential for errors in critical human actions
> Format based on RG 1.206
> Content based on RG 1.206 and SRP
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UAP-HF-08011-22
HFE Design Process Overview (18.1)
The US-APWR HSI design is based on the HSI for
Japanese plants, which has been developed in phases
over the past twenty years
The Japanese HSI was developed following the
NUREG 0711 HFE process
V This included dynamic validation by more than 46 Japanese
operating crews (138 operators)
/ V&V included operability by one RO I one SRO
The US-APWR HFE program reassesses each HFE
program element, with emphasis on changes from
prior experience
/ DCD describes applicability of prior HFE and new activities
specific to US-APWR, including additional dynamic
validation by US operators using a full scope simulator.
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Operating Experience Review (18.2)
LERs and SERs
from operating
fromn
PWRs
i Japan
Fsystems,
design of
HFE/HS
standard Japanese
PWR
[2-1oop/3-1oop conventional
PWR, 4-loop APWR]
I
US-APWR HSI Design
US LERs
(from NUREG/CR-6400
Corrective action
systems, Maintenance
Logs and Operating
Logs from US PWRs
Corrective action
Maintenance
Logs and Operating
Logs from operating
PWRs in Japan
OER process was
used for development
of Japanese HSI.
OER will be expanded
for US-APWR.
US LERs and SERs
(post NUREG/CR-6400)
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Function, Task, Reliability Analysis
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Operating Experience Review
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Functional Require ments Analysis and
Function Allocation
-Task definition
-Function allocation
PRA
j(Human
- Computer)
I
&Qualification
organization
Human System
Interface Design
I
HRA and PRA are
integrated to ensure
human actions are
accurately modeled in the
PRA, and to ensure risk
significant human actions
are given increased
attention during the HFE
design process.
Procedure
Deveill pment
allocation
:Information
Display & control
-Protatyping
i
Human Factors
Verification and Validation
* Validation test
- Static test using mockup
- Dynamic test using full-scope simulator
Design Implementation
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> Necessary number of Reactor Operators (RO) is reduced
from 2 to 1 by reduction of Workload for Operation
> Minimum staff complies with 10 CFR 50.54(m)
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UAP-HF-08011-26
HSI Design (18.7)
Large Display Panel
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HSI desian includes
num
Inventc•ry of Fixed
aPositioi
Minin n indications,
alarms and controls.
i
Operational
Display
PJi .yP F
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Operating
Procedure
Display
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Display
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HSI Design (18.7)
> Safety and Non-safety components can be operated from the same screen
Design
consistency
between nonsafety and safety
VDUs facilitates
operator
transition
between HSI
features.
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UAP-HF-08011-28
Procedures and Training (18.8, 9)
;
>
>
>
_
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Training and procedures encompass the full range of
personnel, functions and systems which may affect
plant safety.
Procedures and training material are developed based
on documented Writer's Guides
Computer based procedures include hot links which
display plant information and controls on adjacent
screens
Design consistency between computer and paper
procedures facilitates operator transition for degraded
HSI conditions.
Procedures are validated through dynamic simulation
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UAP-HF-08011-29
HSI Validation (18.10)
HSI Simulation Facility - Pittsburgh (April 2008)
-Full Scale MCR
-Interactive, full functionality
VDUs
'*High fidelity dynamic plant
model
14ft (4m)
17ft (5m)
Used initially to validate Japanese HSI design by US operators (12/2008)
Used later to validate HSI feature changes for US-APWR (6/2009)
Used to validate final US-APWR HSI, including all displays, alarms,
controls and procedures (ITAAC Closure)
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--
UAP-HF-0801 1-30
Implementation and Performance Monitoring ,-i.
(18.11, 12)
Recurring implementation and changes to the HSI
after validation are in accordance with the Design
Implementation process
,, The design change process is based on a risk assessment
including the risk significance of effected human actions
Human performance is monitored on an ongoing
basis to ensure:
> The HSI does not create human performance problems
> Actual human performance is consistent with plant analysis
assumptions regarding credited manual actions
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> MHI I&C, HSI and Electrical Systems provide significant
advancements to
V improve plant safety and availability
/ reduce operations and maintenance costs
> The systems employ proven designs with many years of
demonstrated reliability
> MHI suggests frequent technical meetings to minimize
misunderstandings and thereby facilitate an efficient
regulatory review process
> MHI invites the NRC staff to visit the following facilities
/ MELCO digital I&C factory (Kobe)
/Gas-turbine generator qualification test facility (North
Carolina)
/HSI simulation facility (Pittsburgh)
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UAP-H F-08011-32
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UAP-HF-ARAI
UAP-HF-08011-32
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 9
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
-
iETIIELI
L
LTD.
UAP-HF-08012
Presenter
(AIP
U.IEAffIJmDUSTRIDEES,
Andrew B. Johnson
Principal Engineer
Mitsubishi Nuclear Energy Systems, Inc.
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UAP-HF-08012-1
Contents
1. Overview of Chapter
/
Title of Chapter
v/ Scope of Chapter
2. Contents of Subsections
3. Design features (for example)
F
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1.
UAP-HF-08012-2
Overview of Chapter
)Title of Chapter
Chapter 9: AUXILIARY SYSTEMS
ýScope of Chapter
This Chapter includes the following Sections and
Attachment:
- 9.1
- 9.2
- 9.3
- 9.4
: Fuel Storage and Handling Systems
: Water Systems
: Process Auxiliaries
: Air Conditioning, Heating, Cooling, and
Ventilation Systems
- 9.5 : Other Auxiliary Systems
- Attachment 9A : Fire Hazard Analysis
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2. Contents of Subsections
>Section 9.1 : Fuel Handling and Storage Systems
Regulatory Guide 1.206
9.1.1 Criticality Safety of Fresh and
Spent Fuel Storage
9.1.2 New and Spent Fuel Storage
9.1.3 Spent Fuel Pool Purification and
Cooling System
9.1.4 Light Load Handling System
(Related to Refueling)
9.1.6 Overhead Heavy Load Handling
System
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US-APWR DCD
9.1.1 Criticality Safety of New and
Spent Fuel Storage
9.1.2 New and Spent Fuel Storage
9.1.3 Spent Fuel Pit Purification and
Cooling System
9.1.4 Light Load Handling System
(Related to Refueling)
9.1.5 Overhead Heavy Load Handling
System
9.1.6 COL Information
9.1.7 References
TIES, LTD.
UAP-HF-08012-4
evink
2. Contents of Subsections
>Section 9.2: Water Systems
Regulatory Guide 1.206
9.2.1 Station Service Water System
9.2.2 Cooling System for Reactor
Auxiliary
9.2.3 [ Reserved ]
9.2.4 Potable & Sanitary Water Systems
9.2.5 Ultimate Heat Sink
9.2.6 Condensate Storage Facilities
US-APWR DCD
9.2.1 Essential Service Water System
9.2.2 Component Cooling Water System
9.2.3 [ Reserved 1
9.2.4 Potable & Sanitary Water Systems
9.2.5 Ultimate Heat Sink
9.2.6 Condensate Storage Facilities
9.2.7 Chilled Water Systems
9.2.8 Turbine Component Cooling Water
SYstem
9.2.9 Non-Essential Service Water
9.2.10 COL informaton
9.2.11 References
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2. Contents of Subsections
>Section 9.3: Process Auxiliaries
Regulatory Guide 1.206
9.3.1 Compressed Air Systems
9.3.2 Process and Postaccident
Sampling Systems
9.3.3 Equipment and Floor Drainage
System
9.3.4 Chemical and Volume Control
System (PWR Only)
9.3.5 Standby Liquid Control System
(BWR Only)
-__ _
.
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US-APWR DCD
9.3.1 Compressed Air and Ga Systems
9.3.2 Process and Postaccident
Sampling Systems
9.3.3 Equipment and Floor Drainage
Systems
9.3A Chemical and Volume Control
System
9.3.5 Not ADplicable for US-APWR
9.3.6 COL Information
9.3.7 References
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2. Contents of Subsections
>Section 9.4: Air Conditioning, Heating, Cooling, and
Ventilation Systems
Regulatory Guide 1.206
9.4.1 Control Room Area Ventilation
System
9.4.2 Spent Fuel Pool Area Ventilation
System
9.4.3 Auxiliary & Radwaste Area
Ventilation System
9.4.4 Turbine Building Area Ventilation
System
9.4.5 Engineered Safety Feature
Ventilation System
US-APWR DCD
9.4.1 Main Control Room Heating.
Ventilation & Conditionina System
9.4.2 Spent Fuel Pool Area Ventilation
System
9.4.3 Auxiliary Building Ventilation
System
9.4.4 Turbine Building Area Ventilation
System
9.4.5 Engineered Safety Feature
Ventilation System
9.4.6 Containment Ventilation System
-9.4.7 COL information
9.4.8 References
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2. Contents of Subsections
>Section 9.5: Other Auxiliary Systems
Regulatory Guide 1.206
9.5.1 Fire Protection Program
9.5.2 Communication System
9.5.3 Lighting System
9.5.4 Diesel Generator (DG) Fuel Oil
Storage & transfer System
9.5.5 DG Cooling Water System
9.5.6 DG Starting Air System
9.5.7 DG Lubrication System
9.5.8 DG Combustion Air Intake &
Exhaust System
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9.6.1 Fire Protection Program
9.5.2 Communication System
9.5.3 Lighting System
9.5.4 Gas Turbine Generator Fuel Oil
Storage & transfer System
9.5.5 Not Applicable for US-APWR
9.5.6 Gas Turbine Generator Starting Air
System
9.5.7 Gas Turbine Lubrication System
9.5.8 GTG Combustion Air Intake and
Exhaust System
9.5.9 COL information
9.4.10 References
LTD.
UAP-HF-08012-8
3. Design Features (for example)
> Support Systems for Safe Shutdown
V' Component Cooling Water Systems (CCWS)
V Essential Service Water Systems (ESWS)
/
HVAC systems, etc.
> Design Features of CCWS & ESWS
/
CCWS and ESWS constitute a safety cooling chain
V 4 Train configuration
/
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UAP-HF-08012-9
3.
Design Features (for example)
);oComponent Cooling Water System
v' 4 safety train configuration
(Each train includes 1 CCWP and 1 CCW HX)
v/ Separated into 2 independent sections
(Each section has 1 CCW surge tank)
,/The other safety components (e.g.;SFP HX) are supplied
with cooling water from 2 of 4 safety trains
CCW SURGE TANK
CCW SURGE TANK
CC
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.UAP-HF-08012-10
MES, LTD.
3. Design Features (for example)
QAýPSW-
•Essential Service Water System
v/ Completely independent 4 train configuration
( Each train includes 1 ESWP )
/ Raw water cooling for the CCW HX and Essential
Chiller Unit
CCW Hx
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ESWP
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Chiller
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US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 10
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
"MLTSUKI•ISHIAV--IlnDU
TRIES. LTD.
UAP-HF-08013
Presenter/Section Leader
Yoshihiro Minami
Engineering Manager
Nuclear Turbine Plant Engineering Section
Water Reactor Engineering Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, LTD.
-
LTD.
EMU-BMSA•
•U1SXR-IES,
H APi-WPAnfl4*A-
Contents
1. Overview of Chapter
Title of Chapter
Scope of Chapter
Overall System Flow Diagram
2. Design Features
Significant Design Features
System Design Features
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UAP-HF-08013-2
1. Overview of Chapter
>Title of Chapter 10
/
STEAM AND POWER CONVERSION SYSTEM
Scope of Chapter
>
This chapter includes the design description of the
systems and the components for power conversion
V" This chapter consists of 4 sections:
,
•
Section 10.1 • Summary Description
*
Section 10.2 Turbine-Generator
*
Section 10.3: Main Steam Supply System
Section 10.4 : Other Features of Steam and Power
Conversion System
This chapter deal with 13 systems in total
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1. Overview of Chapter
QUAJ#-n! k
10.1 Overall System Flow Diagram
CN'<-r->RB
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10.3 Main Steam Supply System:
1
(MSS)
Containment Vessel
Reactor Building
Turbine Building
Auxiliary Building
CV:
CV:
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*
410
2Turbine-Generator (T/G)
10.4.2 Main Condenser
D10.4.1
Turbine
10.4
Gland
Main Condensers
(CS
Evacuation System
Bypa toss
(TBS)
D
10.4.8 Steam Generator
Blowdown System
(SGBDS)
Water Sse
0~~10.4.5'"culat-ing
I
CS
10.4,3 Gland Seal System (GSS)
10.4.6 Condensate polishing
System (CPS)
10.4.7 Condensate and Feedwater System (CFS)
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Design Features
10.1 Significant Design Features
Rated NSSS power (MWt)
Steam Generator Outlet Press. (psig)
4,466
957
Quantity of Steam Generator (SG)
Total steam flow rate from SG (Ib/hr)
4
20,200,000
Steam Turbine Rating
Type of Steam Turbine (-)
Rotating Speed (rpm)
TC6F
1,800
Generator Output (MWe)
Exhaust Pressure (inHga)
1,700
1.5
Generator Rating
1,900
0.9
Capacity (MVA)
Power Factor (-)
14
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2. Design Features
10.2 Turbine-Generator (T/G)
Low Pressure Turbines (LPTs)
High Pressure Turbine (HPT)
Generator
Moisture Separator/Reheater (MS/R)
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2. Design Features
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S.
10.2 Turbine-Generator (TIG)
,/
The T/G is non safety-related system
V The T/G could be a potential source of a high-energy
turbine missile, which could cause damage to safetyrelated equipment or systems
/
Turbine and control/protection system are to be designed
so that probability of turbine missile generation probability
satisfies the requirement of SRP (less than 1 x 10-5 per
year assuming proper inspection and test frequency)
V The orientation of the T/G is such that a high-energy
missile to be directed at an approximately 90 degree
angle away from the safety-related structures
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2. Design Features
> 10.3 Main Steam Supply
System (MSS)
V The MSS is to transport steam
from the SGs to the HPT and to
the MS/R
V The MSS is provided with safety-
related main steam isolation
valves (MSIVs) and main steam
bypass isolation valves (MSBIVs)
in each main steam line for the
purpose of:
* Isolating the secondary side of
the SGs to prevent the
uncontrolled blowdown of
more than one SG
* Isolating non safety-related
portions of the system
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2. Design Features
10.4.1 Main Condenser
V/The main condenser is non. safety-related system
V The main condenser functions to condensate and
deaerate the exhaust steam from the main turbine and
provide a heat sink for the turbine bypass system
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2. Design Features
10.4.2 Main Condenser Evacuation System
(MCES)
The iCES is non safety-related system.
The iCES removes noncondensable gases from the
main condenser during plant startup and normal
operation
/
V The iCES establishes and maintains a vacuum in the.
main condenser
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UAP-HF-08013-10
2. Design Features
10.4.3 Gland Seal system (GSS)
V The GSS is non safety-related system.
GSS prevents air leakage into and steam leakage
out of the casing of the steam turbine
/The
/
Sealing steam is supplied to the turbine shaft from
either the Auxiliary Steam Supply System (ASSS) or
the MSS
system returns the steam-air mixture from the
turbine glands to the gland steam condenser and
exhausts non-condensable gases into the atmosphere
/The
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UAP-HF-08013-11
APH-831I
UAP-HF-08013-11
2. Design Features
10.4.4 Turbine Bypass system (TBS)
$ The TBS is non safety-related system.
v/ The TBS is part of the MSS and provides capability to
send the main steam flow from the SGs to the main
condenser bypassing the main turbine
V The TBS is designed to sustain a 100% load rejection
without reactor trip, and not requiring actuation of the
main steam relief valves, main steam safety valves and
pressurizer safety valves
D
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UAP-HF-08013-12
2. Design Features
10.4.5 Circulating Water System (CWS)
/ The CWS is non safety-related system.
/ The CWS supplies cooling water to remove the heat
from the main condenser under various plant operating
conditions and site environmental conditions
V The CWS removes the plant heat during startup,
normal operation, shutdown, transient condition, or
turbine trip
LMI_.7rSMJSVJJHEAVY-MDUSTRIES, LTD.
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UAP-HF-08013-13
UAP-HF-0801 3-13
2. Design Features
QA PSý4f
> 10.4.6 Condensate Polishing System
(CPS)
/ The CPS is non safety-related system.
,/The CPS is designed to remove dissolved ionic solids
and impurities from the condensate and assists in the
removal of corrosion products
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2. Design Features
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10.4.7 Condensate and
Feedwater System (CFS)
,/The CFS provides feedwater at the
required temperature, pressure,
and flow rate to the SGs
/ The safety-related function of the
CFS is to provide containment and
feedwater isolation following a
design basis accident
/ The system provides main
feedwater isolation valves (MFIVs)
in the main feedwater lines
MFIVs close to limit the mass
and energy release to the
containment
/The
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UAP-HF-08013-15
2. Design Features
10.4.8 Steam Generator Blowdown
System (SGBDS)
I
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/ The SGBDS assists in maintaining
secondary side water chemistry
within acceptable limits during normal
operation and during anticipated
operational occurrences due to main F ' '
condenser tube leakage or primary to L .secondary
steam generator tube
leakage
[,: :•II
i
V The SGBDS has a safety-related
function to isolate the secondary side
of the SGs using two isolation valves
in series in the blowdown line from
eachSG
v' This provides a heat sink for a safe
shutdown or to mitigate the
consequences of a design basis
accident
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2. Design Features
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> 10.4.9 Emergency Feedwater System
(EFWS)
," The EFWS is a safety-related system
/ The EFWS is designed to supply
feedwater to the SGs and remove
reactor core decay heat following
transient conditions or postulated
accidents such as:
" Reactor trip
" Loss of offsite power (LOOP)
• Loss of main feedwater
" Feedwater line break (FLB)
* Main steam line break (MSLB)
" The EFWS consists of two motor-driven
emergency feedwater (EFW) pumps,
two turbine-driven EFW pumps,
emergency feedwater pits and other
.necessary equipment
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2.
Design Features
> 10.4.10 Secondary Side Chemical Injection
System (SCIS)
, The SCIS is non safety-related system.
/The SCIS is designed to maintain a noncorrosive
condition within the secondary loop
,/ Noncorrosive condition is maintained by controlling pH
and dissolved oxygen content in the secondary side
by:
* Maintaining alkaline pH by ammonia injection
" Scavenging dissolved oxygen with hydrazine
injection
UAP-HF-08013-18
_MIE.SýUBI$H I-HE-•/-Y-INDRU STRIES, LTD.
2. Design Features
•A4
10.4.11 Auxiliary Steam Supply System
(ASSS)
V"The ASSS is non safety-related system.
v/ The ASSS is designed to provide the steam required
for plant use during plant startup, shutdown, and
normal operation
V Steam is supplied from either the auxiliary boiler or the
steam converter
L*L1ArURJSjkf1= jEAV_
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LTD.
UAP-HF-08013-19
UAP-HF-0801 3-19
Summary
Chapter 10 deals with the steam and power
conversion system
The steam and power conversion system is
designed to remove the heat energy from the
reactor coolant system and to convert it to
electrical energy in a safe manner
The turbine and control/protection systems are
designed so that the probability of turbine missile
is less than the number specified in SRP
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~MUISUB ISHI-KE-AVY-INDUSTRIES. LTD.
UAP-HF-08013-20
UAP-HF-ORO1 ~-2fl
UAP-HF-08013-20
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 11 (Dose Evaluation)
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
SITSUISI= AVY-INDUSTRIES,
LTD.
Presenter
UAP-HF-08015
APW
Hiromasa Nishino
Engineering Manager
Radiation Safety Engineering Section
Reactor Safety Engineering Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, LTD.
L-MijX-$jQ1%1j5 "I= J_
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S
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~.MI.1S.IA~JS!.B~HEAV.Y4NDUSTRIES
LTD.
UAP-HF-08015-1
UAP-HF-0801 5-1
Contents
1. Overview of Chapter
/
Title of Chapter
v' Scope of Chapter
2. Design Features
3. Dose Evaluation Methods, Criteria
and Results
4. Summary
I SUB ISHIEA-Y-INDU
RIE S, LTD.
UAP-HF-08015-2
1. Overview of Chapter
>
Title of Chapter
Chapter 11: Radioactive Waste Management
>
Scope of Chapter
This chapter includes following items;
o
Source Term
* Liquid Waste Management System (LWMS)
o
Gaseous Waste Management System (GWMS)
Solid Waste Management System (as presented
in "Detail of FSAR Tier 2 : Chapter 11(System)")
Process Effluent Radiation Monitoring and
Sampling System (ditto)
I
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~-MI.TSUBISHU-HEAV-Y-INDUSTRIES.
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UAP-HF-08015-3
UAP-HF-0801 5-3
1.
Overview of Chapter (Cont'd)
QP
> Scope of Chapter (Cont'd)
Each item described above includes subitems as
follows;
Source Term
v/ Design Basis Reactor Coolant and Secondary
Coolant Activity
,,'Realistic Reactor Coolant and Secondary Coolant
Activity
Notes : Fission products and activation products are considered.
M
a!SEUSBISHHEvY-IDUSTREES,
1.
LTD.
UAP-HF-08015-4
Overview of Chapter (Cont'd)
AW7
> Scope of Chapter (Cont'd)
• LWMS (Radioactive Effluent Releases)
/ Radioactive Effluent and Dose Calculation in
Normal Operation
v' Radioactive Release due to Liquid Containing
Tank Failure
GWMS (Radioactive Effluent Releases)
/ Radioactive Release and Dose Calculation in
Normal Operation
/Radioactive Release and Dose Calculation due to
GWMS Leak or Failure
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UAP-HF-08015-5
UAP-HF-0801 5-5
2. Design Features
"
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Chapter 11 presents information on source terms
of radioactive material generated within reactor
core and released via LWMS and GWMS.
* Two source term models are utilized to calculate
the radionuclide concentration in the reactor
coolant and secondary coolant.
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2. Design Features (Cont'd)
AP
Design basis source term
(for shielding design)
" Fuel defect:1 %
" Mass balance equations described in DCD are
used to calculate each nuclide activity.
> Realistic source term
(for dose evaluation during normal operation)
* Based on ANSI/ANS-18.1-1999(*)
* PWR-GALE Code is used to calculate realistic
source term and released activity during normal
operation.
(*)equivalent to approximately 0.2% of fuel defect
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1%
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UAP-HF-08015-7
5-7
UAP-HF-O8O1
3. Dose Evaluation Methods, Criteria and
Results
f-
_____
Evaluation Item
Individual dose during
normal operation
Evaluation Method
Radioactive Releases: PWR-GALE Code
Dose Evaluation (Liquid) : LADTAPII Code
Dose due to GWMS
leak or failure
Dose Evaluation (Gaseous) : GASPARII Code
Branch Technical Position 11-5
Considered operational mode of US-APWR
Radioactive Effluent
Releases due to liquid
Branch Technical Position 11-6
NUREG-0133 Appendix A (RATAF Code)
- containing tank
failure
Each evaluation is performed using assumed
conservative site characteristics.
UAP-HF-08015-8
_MI=T.SUBISHI-_HEAV.Y-!ND UiS•RI ES, LTD.
3. Dose Evaluation Methods, Criteria and
Results (Cont'd)
Evaluation Item
Criteria
Individual dose during
normal operation
10 CFR 50 Appendix I
Liquid
Results
Total body
Organ
3 mrem/y
10 mrem/y
Gaseous (Noble gases)
Gamma dose in air
10 mrad/y
Beta dose in air
20 mrad/y
Total body
5 mrem/y
Skin
15 mrem/y
Gaseous (Iodine, Particulates)
Organ
15 mrem/y
Dose due to GWMS leak
or failure
Branch Technical Position 11-5
100 mrem
Radioactive Effluent
Releases due to liquid containing tank failure
10 CFR 20 Appendix B Table 2 Col.2
(Summation of fractions of concentration
limit is equal to or less than 1.0)
*Child **Child's Liver
Child's bone
****
L-STiRIES. LTD.
1.98 mrem/y*
2.54 mrem/y**
0.210 mrad/y
1.62 mrad/y
0.134 mrem/y
1.26 mrem/y
10.2 mrem/y***
46 mrem
0.22
Summation of Fractions of Concentration
UAP-HF-08015-9
4. Summary
Source terms for radiation protection design
and dose evaluation were evaluated using
two source term models, i.e. design basis
source term and realistic source term.
Using assumed conservative site
characteristics, dose evaluations were
performed according to standard methods
in the U.S. and complied with dose criteria.
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[IAP-HF-ORt31
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UAP-HF-08015-10
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 12
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
_=
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U=AI'J
*PJMIC7EPCTDE
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Presenter
I IAP-HF-flAfltA
UAP-HF-08016
(!APW0
Hiromasa Nishino
Engineering Manager.
Radiation Safety Engineering Section
Reactor Safety Engineering Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, LTD.
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~MEISUBISHI-HEAV-Y-INDUS.T~RIES. LTD.
UAP-HF-08016-1
UAP-HF-08016-1
Contents
1. Overview of Chapter
Title of Chapter
/
/ Scope of Chapter
2. Design Features
3. Summary
~MLSUBSHIIEV-YINDSTIES,
I.
LTD.
U/r-lrM-USUi b-/.
Overview of Chapter
STitle of Chapter
Chapter 12: Radiation Protection
> Scope of Chapter
This chapter includes following items;
/ Considerations for ALARA*
v/ Radiation Sources
v/ Radiation Protection Design Features
*ALARA: As Low As Reasonably Achievable
LMI:IjSQBtSHj-KEAV-XjP
- JWU_STWIES, LTD.
LTD.
UAP-HF-08016-3
UAP-HF-0801 6-3
1.
Overview of Chapter (Cont'd)
> Scope of Chapter (Cont'd)
Each item described above includes subitems as follows;
Considerations for ALARA
v/ Policy Considerations
v/ Design Considerations
,/ Operational Considerations
"
-
COL items
Radiation Sources
Sources
/ Airborne Sources
/Contained
_. MSUBQISH4I EAV.-INDHUS1RIES, LTD.
UAP-HF-08016-4
1. Overview of Chapter (Cont'd)
> Scope of Chapter (Cont'd)
Radiation Protection Design Features
v/" Plant Design Features for ALARA
'./ Shielding
/v Ventilation
/Area Radiation and Airborne Radioactivity
Monitoring Instrumentation
/ Dose Assessment
Note: Operational Radiation Protection Program
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~~METSUBISHI~HEAVY-ENDUSTRIES.
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COL Item
UAP-HF-08016-5
6-5
UAP-HF-OBO1
UAP-HF-08016-5
2. Design Features
A
> Ensuring that Occupational Radiation
exposures are ALARA
*Policy Considerations
* Design Policies: Design by nuclear engineers with ALARA
philosophy and system to cooperate plant experience
* Operation Policies: Comply with RG 1.8,8.8&8.10 - COL Item
*Design Considerations
Equipment and Facility Layout are designed to minimize the
personnel time spent in radiation areas and to minimize the
radiation levels in routinely occupied plant areas
_M/.TSUBIISI-HEAV-Y-!MDU $gTkR
I ES, LTD.
UAP-HF-08016-6
2. Design Features (Cont'd)
> Radiation Sources
*Sources for Full-Power Operation
-Contained Sources : 1% Fuel defect considered
- Airborne Sources : Constant leakage from equipments to atmosphere
considered
*Sources for Shutdown
* Reactor Core: Specific Power of 32.1 MW/MTU and two cycles operation
considered
* Spent Fuel : Specific Power of 32.1 MW/MTU and Burn-up of 62
GWD/MTU considered
* Incore Flux Thimbles : Activated Cobalt-60 considered
*Sources for Design-Basis Accident
* Fission Products released into the containment based on RG 1.183 considered
Aý= M-12TSLUERSH1-HEAV;Y=tjWqU.ST4t1ES, LTD.
I~UBISNLHEAV~Y~INDUSXRIES. LTD.
UAP-HF-08016-7
UAP-HF-0801 6-7
2. Design Features (Contd)
k,4AfPWRV
>Radiation Protection Design Feature
*Facility Design Features " All equipment is designed to ensure
the Occupational Radiation Exposures ALARA
*Shielding Design : designed to be in
* Compliance with 10 CFR 20 under normal operation/shutdown
* Compliance with 10 CFR 50 Appendix A and NUREG-0737 under
Design-Basis Accident (Main Control Room)
*Ventilation Design & Area Radiation and Airborne Radioactivity
Monitoring Instrumentation Design : considered for ALARA
*Dose Assessment
* Occupational Exposure: about 70 Person-rem/year
* Post-Accident Actions : Radiation Exposures in Post- Accident
Sampling are compliance with 10 CFR 50.34 (f)(2)(viii)
* Radiation Exposures at Site Boundary:
- Direct Radiation : negligible
- Dose due to Airborne Radioactivity : given in Chapter 11
~I~1I5WjEAAI~.NDISTRESLTD.
UAP-HF-08016-8
2. Design Features (Con'd)A
)Radiation Zones for Shielding Design and Radiation Control
Zone
one
I
III
Maximum
Description
Dose Rate
• 0.25 mrem/h
Controlled area, unlimited occupancy
II
III
1 mrem/h
Restricted area, limited occupancy
2.5 mrem/h
Restricted area, limited occupancy
IV
15 mrem/h
Restricted area, limited occupancy
V
100 mrem/h
Restricted area, limited occupancy
VI
1 rem/h
VII
10 rem/h
High radiation sources. Restricted area, limited occupancy for
very short periods. Access controlled as stated in the Technical
Specifications.
Same as Zone VI above
VIII
100 rem/h
Same as Zone VI above
IX
500 rad/h
Same as Zone VI above
X
> 500 rad/h
Very high radiation sources. Restricted area, very limited
occupancy for the shortest periods. Access controlled as stated in
the Technical Specifications.
A2AMrTSUB1j5j"1=H.ýAVX_1NQWSTk1ES, LTD.
ESUBISHI-HEAV-Y- ENDUSi RUES, LTD.
~MI
UAP-HF-08016-9
UAP-HF-08016-9
3. Summary
Policy Considerations,Design Considerations,
Radiation Sources and Radiation ProtectionDesign
Featuresto ensure that OccupationalExposures
are ALARA are describedin chapter 12.
Radiation ProtectionDesign complies with 10 CFR
20 and 10 CFR 50 for Normal Operation/Shutdown
and Post-accidentActions
Dose Assessment for OccupationalExposures and
post-accidentactions meet NRC's general
requirements and/or 10 CFR 50.34
LMISU
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V-Y-IIJUSXTJ.IES, LTD.
UAP-HF-08016-10
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 13
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
-ES, LTD.
UAP-HF-08017
Presenter
Atsushi Kumaki
Engineering Manager
APWR Promoting Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, LTD.
________________MIMES,______
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Contents
1. Overview of Chapter
v' Title of Chapter
v Scope of Chapter
2. Topics of Section
3. Summary
IES, LTD.
UAP-HF-08017-2
1. Overview of Chapter
STitle of Chapter
Chapter 13: CONDUCT OF OPERATION
C-
> Scope of Chapter
This chapter provide information
relating to the preparations and plans
for the design, construction, and
operation
P U5ST"IES, LTD.
UAP-HF-08017-3
2. Topics of Section
> 13.1 Organizational Structure of Applicant
'(
This section makes clear the COL Applicant's
responsibility to describe;
* management and technical support
organization,
" operating organization, and
* Qualification of Nuclear Power Plant
Personnel
NE•UjjI-FlA-Y-INRU5TR
ES, LTD.
UAP-HF-08017-4
2. Topics of Section
o 13.2 Training
V The development of training programs is the
responsibility of the COL Applicant
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2. Topics of Section
13.3 Emergency Planning
v'
This section provides design features
necessary for emergency planning, e.g.;
* technical support center (TSC),
* emergency operations facility (EOF),
* emergency response data system (ERDS),
• data communication system,
* safety parameter display system, and
* post accident monitoring system
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UAP-HF-08017-6
2. Topics of Section
13.4 Operational Program Implementation
/ The development of operational program
implementation is the responsibility of the COL
Applicant
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2. Topics of Section
13.5 Plant Procedures
v" This section makes clear the COL Applicant's
responsibility to develop;
* administrative procedures, and
" operation and maintenance procedures
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UAP-HF-08017-8
2. Topics of Section
13.6 Security
v' This section makes clear the Applicant's
responsibility to develop;
* security assessment,
• plant overall security plan,
0 implementation schedule for the security
program, and
0 proposed ITAAC for physical security hardware
V A security safeguards report will identify vital areas
and vital equipment and other physical protection
information for US-APWR standard design
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2. Topics of Section
13.7 Fitness for Duty
/ The development of the fitness-for-duty
program is the responsibility of the COL
Applicant
1E5••,
LTD.
UAP-HF-08017-10
3. Summary
> Chapter 13 provides information relating to the
preparations and plans for the design,
construction, and operation of the US-APWR
plant.
> The purpose of Chapter 13 is to provide
adequate assurance that the COL Applicant
establishes and maintains a staff of adequate
size and technical competence and that
operating plans to protect the public health and
safety.
IEA
_
I UTIES,R
LTD.
UAP-HF-08017-11
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 14
January 15, 16, 2008
Mitsubishi Heavy Industries, Ltd.
IIAfl
FA Uff#"
PUA
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LAr'-r-1r-uou
Presenter
1O
(fAPS4
Atsushi Kumaki
Engineering Manager
APWR Promoting Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, Ltd.
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-1
Contents
AP
1. Overview of Chapter
v1 Title of Chapter
v- Scope of Chapter
2. Chapter 14 Contents
3. Summary
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LTD.
UAP-HF-08018-2
di54
Aý
1. Overview of Chapter
Title of Chapter
Chapter 14: VERIFICATION PROGRAMS
SScope of Chapter
This chapter consists of
1) Initial Test Program Part (14.1 & 14.2)
(Administrative Control & Test Abstracts)
2) ITAAC Screening Part (14.3)
3) Supplemental Information (Appendix 14A)
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1. Scope of Chapters and Interfaces
2.1 Initial Test Program
(Section 14.1 &14.2)
2.2 ITAAC Screening
ITAAC Selection
Methodology
Cross Reference
of Key Design
14.2
L7
Subpart
(Section 14.3)
S[RG 1.206
Administrative--R
Test Abstract
P
1.687
between Tier 2
and Tier I
Appedix14A
I:
2.3 Supplemental
-Tir
O -he Tier
2
'Other
-:
,Information
m
Chapters
--
Regulatory
Guidance
Ch. 14 Scope
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2. Tier 2 Chapter 14 Contents
"Verification Programs"
2.1 Initial Test Program Part (14.1 & 14.2)
2.2 ITAAC Screening Part (14.3)
2.3 Supplemental Information (Appendix 14A)
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UAP-HF-08018-5
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.54
P
2.1 Initial Test Program Part
'i
Section 14.2 consists of
> Administrative Control Subpart
This part addresses general commitment of the
administrative control.
Site-specific administrative control is described
in the COLA phase.
> Test Abstract Subpart
Most of the test abstracts is addressed in the
DCD. Some of the site-specific test (e.g.
personnel monitors and radiation survey
instruments) will be added in the COLA.
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2.1 Contents Example
>
A!PS:W4
Test abstracts are developed based on the
MHI's initial test experience and U.S. regulatory
guidance (including the past FSAR).
14.2.12.1.5 Pressurizer Relief Tank Preoperational Test
A. Objectives
1. To demonstrate that design pressurizer relief tank spray flow
2. To demonstrate the filling and draining operation of
B. Prerequisites
.
1. Required construction testing is co
2. The containment vess I react
I
t
...... is available to the drain .....
C Tes .... tho
1. Witest
2. While0•
__
r
..
s re ....... the required spray flow is pumped
to the pressurizer relief tank.
the nitrogen pressurization system operation is observed.
D. Acceptance Criteria
1. The required spray flow is obtained as designed (see Subsection 5.4.11)
2. The pressurizer system ......
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UAP-HF-08018-7
UAP-HF-0801 8-7
2.2 ITAAC Screening Part
Section 14.3 consists of
; ITAAC Selection Methodology
All selection methodology (including for
Emergency Planning ITAAC and Physical
Security ITAAC) is provided in the DCD.
>Cross Reference Table between Tier
2 Key Design Features and Tier 1
Description.
The significant parameters and key design
features in Tier 2 are listed with the applicable
Tier 1 description and section numbers.
!,I.S
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UAP-HF-08018-8
IES, LTD.
US,
2.2 Contents Example
The cross-connection clearly shows how
and where most significant key design are
addressed in Tier I and Tier 2.
Tier I Ref.
Subsection
2.7.1.2.1
Key Design Features
The valves close within the
receipt of an actuation si
The main steam is I
D
within 5 c
Thesrd
Thees27.12-4
Table 2.7.1.2-4
L*UWBtS#,ff
I
Tier 2
D ILocation
fter
~SIVs) close
capacities of the MSSVs
0l,000 Ib/hr....
The flow restrictor within the SG.......
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Subsection
10.3.2.3.4
Subsection
10.3.2
Table 10.3.2-2
Subsection
15.1.5.2
UAP-HF-08018-9
UAP-HF-0801 8-9
2.3 Supplemental Information
1::::::A
ýPS
•Comparison Table with RG 1.68
/ Each item of the RG 1.68 Appendix A is
listed with applicable test abstracts.
/ Exceptions from the regulatory
guidance are clearly specified with the
justification.
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2.3 Contents Example
Comparison Table Example in Appendix 14A
-/Thiscross-connection clearly shows the conformance
with the regulatory guide.
Section
Number
RG 1.68
Appendix A
i
1.h.A7•
Typical Test
i
14.2.12.1.57
m ulator Testing
Safe•
D
•~tr
Storage
System
14.2.12.1.59
\•,•ational Test
I .h.(8)
i
IWot applicable
This system does not have an ESF
function in the US-APWR.
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MI~UBISHI~HEAV~W~4NDUSJRIES. LTD.
UAP-H F-08018-11
UAP-HF-08018-1
UAP-HF-08018-11I
3. Summary
(A#4
)Chapter 14 contents completely
conform to RG 1.206.
.Cross-connection between RG 1.68
and individual test abstracts are
available for the reviewer's
convenience.
)These contents provide sufficient
information for NRC's review.
IX-S-U-
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UAV-HI--M5UItS-Id
14
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier 2: Chapter 15
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
lIr-SUIH 1
WE A--N DuIlE
S, LTD.
Presenter
UAP-HF-08019,
UAP-HF-08019.
!AP&*'
Keith Paulson
Senior Technical Manager and Licensing Manager
Mitsubishi Nuclear Energy Systems, Inc.
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LIAP-HF-08019-1
P-H F-08019 -1
Contents
1. Overview of Chapter 15
2. Design Features Related to Transient and
Accident Analyses
3. Selection of Design Basis Events and
Acceptance Criteria
4. Event Categorization and Computer Codes
Used
5. Analysis Methods
6. Analysis Results
7. Summary
EI.SUBISHI_.H EAV-Y-•.IND•UU$T•IES,
LTD.
UAP-HF-08019-2
1. Overview of Chapter 15
STitle of Chapter
Chapter 15: Transient and Accident Analyses
> Scope of Chapter
Transient and Accident analyses reported in the
Design Control Document (DCD) include eight
(8) categories of events to comply with the
Regulatory Guide (RG) 1.206 and Standard
Review Plan (SRP) NUREG-0800
MITSU BISHI-HEVVY-!NDUSRES,
LTD.
UAP-HF-08019-3
2. Design Features Related to Safety Analysis
-
US-APWR Plant ParameterSummary
V' Larger core thermal output with improved efficiency
V Enhanced thermal margins due to the lower average linear
heat rate
US Current
4 Loop Plant
Features
Core thermal output (MMt)
4,451
3, 565
4
4
257
193
17x 17
17x 17
Active fuel length (ft)
14
12
Average linear heat rate (kW/ft)
4.6
5 7
Centrifugal
Centrifugal
U-Tube
U-Tube
Number of loops SGs and RCPs
Number of fuel assemblies
Fuel rod lattice
Reactor coolant pump type
Steam generatortype
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2. Design Features Related to Safety Analysis (Cont'djfS
US-APWR Design Features
V Very similar to current PWRs in the US
/ Design Features and the Effects on Safety Analyses
Featumres
I
Neutron Reflector
Simplified core lower
Effects on Safety Analyses
Neutron Reflector is explicitly modeled in LOCA analyses
Negligible change in neutron kinetics
Core inlet mixing among loops approximately the same
plenum
Pressurizer
Largersteam space moderates pressure transients
Steam generator
Smaller U-tube diameter improves transientperformance in case
of SGTR*'
ECCS and EFWS*2
4 independent trains with one pump per train
Diverse actuation
system
Satisfies design requirements to cope with A TWS*3
Advanced
Accumulator
Characteristicsof Advanced Accumulator is modeled in LOCA
analyses
Not expected to actuate during Non-LOCA events
*2 EFWS -Emergency FeedwaterSystem
*1 SGTR -Steam Generator Tube Rupture;
*3A TWS -Anticipated Transients Without Scram
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.AVY
MDUSTRIES, LTD.
UAP-HF-08019-5
1
3. Selection of Design Basis Events and
Acceptance Criteria
> Design Basis Events
V Basic design of the US-APWR is the same as the current PWRs
in the U.S. from the viewpoint of
- primary and secondary system configurations
- thermal hydraulic characteristics and main plant parameters
- fuel properties
- core kinetics
- reactor control and protection system functional design
/
The US-APWR design does not introduce any new initiating
events for safety evaluation.
/ All transientsand accidents in NRC StandardReview Plan (SRP)
Chapter 15, applicable to PWRs, are included
$ 8 categories based on the causes of transientsconsistent with
the SRP
> Acceptance criteria
V SRP Acceptance Criteriaare applied for US-APWR analyses
-MITSMUBI
AHHE
11-
V
UAP-HF-08019-6
UAV
SIES, LTD.
4. Event Categorization and Computer Codes Used
> SRP Chapter 15 Events, Classification.C mpuIter Codes
=
Section
Events
15.1.1
Decrease in feedwater temperature
AOO
MARVEL-M
15.1.2
Increase in feedwater flow
AOO
MARVEL-M
15.1.3
Increase in steam flow
AOO
MARVEL-M
15.1.4
Inadvertent opening of a steam generator relief
or safety valve
AOO
MARVEL-M, ANC, VIPRE-0IM
15.1.5
Steam system piping failures - Minor I Major
AOO I PA
MARVEL-M, ANC, VIPRE-01M
15.2.1
Loss of external electrical load
AOO
MARVEL-M
15.2.2
Turbine trip
AOO
Bounded by loss of load
15.2.3
Loss of condenser vacuum and other events
resulting in turbine trip
AOO
Bounded by loss of load
15.2.4
Inadvertent closure of main steam isolation
valves
AOO
Bounded by loss of load
15.2.5
Steam pressure regulator malfunction or failure
that results in decreasing steam flow
AOO
No steam pressure
regulators in the.US-APWR
whose malfunction or failure
could result in a steam flow
transient.
15.2.6
Loss of non-emergency AC power to the station
auxiliaries
AOO
MARVEL-M
15.2.7
Loss of normal feedwater flow
AOO
MARVEL-M
15.2.8
Feedwater system pipe break - Minor I Major
AOO I PA
MARVEL-M
EYQ
IR~S.uUA~ii~
Category
smmnisenc
UTE
I ivn
Computer Code(s) Utilized
I AP..w;..flpl4Q.7
4. Event Categorization and Computer Codes Used
(Cont'd)
.. SR
/-us
Chapter15 Events, Classification, Computer Codes
Section
Events
15.3.1.1
Partial loss of forced reactor coolant flow
AOO
MARVEL-M, VIPRE-01M
15.3.1.2
Complete loss of forced reactor coolant flow
AOO
MARVEL-M, VIPRE-01M
15.3.2
Flow controller malfunctions (not applicable to
US-APWR)
15.3.3
Reactor coolant pump rotor seizure
PA
MARVEL-M, VIPRE-01M
15.3.4
Reactor coolant pump shaft break
PA
Bounded by rotor seizure
15.4.1
Uncontrolled RCCA bank withdrawal from a
subcritical or low power startup condition
AOO
TWINKLE-M, VIPRE-01M,
MARVEL-M
15.4.2
Uncontrolled RCCA bank withdrawal at power
AOO
MARVEL-M
15.4.3
RCCA misalignment
15.4.4
Startup of an inactive reactor coolant pump at an
.incorrect temperature
15.4.5
Malfunction / Failure of flow controller in BWR
recirculation loop
-
15.4.6
CVCS malfunction that results in a decrease in
boron concentration in the reactor coolant
AOO
15.4.7
Inadvertent loading and operation with fuel
assembly in improper location
PA
ANC
15.4.8
Spectrum of RCCA ejection accidents
PA
TWINKLE-M, VIPRE-01M,
MARVEL-M
Category
-
Computer Code(s) Utilized
N/A - BWR Event
AOO I PA
MARVEL-M, VIPRE-01M
AOO
N-1 loop operation not
allowed
N/A - BWR Event
Evaluation without computer
code
L_ ISuBJAHiHE AV-Y-iNDUS T IES, LTD.
UAP-HF-08019-8
4. Event Categorization and Computer Codes Used
r•J
(Cont'd)
SRP Chapter 15 Events, Classification, Computer Codes
Section
Event
15.5.1
Inadvertent actuation of the emergency core
cooling system during power operation
AOO
N/A - shut off head of the SI
pump is below nominal
operating pressure
15.5.2
CVCS malfunction that increases reactor coolant
inventory
AOO
MARVEL-M
15.6.1
Inadvertent opening of a pressure relief valve
AOO
MARVEL-M
15.6.2
Radiological Consequences of the Failure of
Small Lines Carrying Primary Coolant Outside
Containment
AOO
RADTRAD
15.6.3
Steam generator tube rupture
PA
MARVEL-M
15.6.4
Radiological Consequences of Main Steam Line
Failure Outside Containment (BWR)
-
15.6.5
Loss-of-Coolant-Accidents Resulting from
Spectrum of Postulated Piping Breaks Within the
Reactor Coolant Pressure Boundary
PA
15.7
Radioactive Release from a Subsystem or
Component
15.8
Anticipated Transient Without Scram
L_
U$_Tk1ES, LTD.
LT$UBISHI~HEAV~V~-INPiJSNRIES, LTD.
Category
AOOIPA
N/A
Computer Code(s) Utilized
N/A- BWR Event
WCOBRA/TRAC, HOTSPOT,
M-RELAP5
RADTRAD
N/A
UAP-HF-08019-9
9-9
UAP-HF-0801
UAP-HF-08019-9
US
5. Analysis Methods
>LOCA
/ Large Break LOCA
WCOBRAITRAC code with the ASTRUM methodology
is implemented
Applicability of this methodology for US-APWR is
submitted in Topical Report entitled "Large Break
LOCA Code Applicability Report for US-APWR"
(MUAP-07011-P(RO), July 2007) and is under review
/ Small Break LOCA
M-RELAP5 code which incorporates Appendix-K
requirements is used
This methodology is submitted in Topical Report
entitled "Small Break LOCA Methodology for USAPWR" (MUAP-07013-P(RO), July 2007) and is under
review
Plant Sensitivity analyses are provided in Technical
Report (MUAP-07025-P(RO), December 2007)
-i._TSUSHI-HEAV-Y-!!-I•NDTUSI ES, LTD.
UAP-HF-08019-10
5. Analysis Methods (Cont'd)
Non-LOCA
/MARVEL-M
':1
code and TWINKLE-M code are
applied
/TWINKLE-M 3-0 neutron kinetics
methodology is used for the Rod Ejection
Accident analysis from HZP condition
,/Applicability of the methodology for
US-APWR is submitted in Topical Report
entitled "Non-LOCA Methodology", (MUAP07010-P (RO), July 2007) and is under review
III.-T.SUBISHI-H&WV4--•ND-USKTRIES,
LTD.
UAP-HF-08019-11
5. Analysis Methods (Cont'd)
!APS4
>Radiological Consequence Analyses
-/Alternative source term is used
,/Methodologies which consider decay, removal
and transport of radioactivity based on plant
design are applied and equivalent to current
U.S. PWR
NJ
v/ RADTRAD code, approved by NRC, is used
41MVKSUBLSHI-EkV-Y-M
$ISTRIES,LTD.
UAP-HF-08019-12
6. Analysis Results
1.
2.
3.
The transient and accident response of the US-APWR is
similar to that of current PWRs in the US
All analysis results satisfy the SRP Acceptance Criteria
Large thermal margin due to the lower average linear heat
rate greatly enhances the safety margins
e No DNB forAOOs
> Minimal fuel failure and radiological consequences for PAs
>
No PCMI fuel failure for Rod Ejection Accident
4. Enhanced ECCS performance
>
Large PCT margin for LOCA
>
Small increase in cladding temperature due to loop seal
formation.
LMI-T-SUWISHI-HEAV-Y-_INDU.ATRIES,
MIThUBISHLHEAVZY-INPMSTRIES.
LTD.
LTD.
UAP-HF-08019-13
UAP-HF-O8O1 9-13
6. Analysis Results (Cont'd)
111ýw-
AQOs (Anticipated Operational Occurrences)
" Minimum DNBR remains above the 95/95 limit and no fuel failures
are predicted
V"RCS pressure and main steam system pressure remain well below
110% of respective system design pressures
V All AQOs do not generate any other fault that may lead to a
postulated accident
> PAs (Postulated Accidents)
V Minimum DNBR remains above the 95/95 limit for most PAs.
If the minimum DNBR falls below the limit, the acceptance criteria
in 10 CFR 50.46 are satisfied
V RCS pressure and the main steam system pressure remain below
acceptable design limits.
V All PAs do not cause any consequential loss of required functions
of systems needed to cope with the fault.
V Resultant doses are well within the guideline values specified in
10 CFR 50.34
"IiICSAJISHIJ4iE-AV2VY-.INRtU5T-RIES,
LTD.
6. Analysis Results (Cont'd)
UAP-HF-0801 9-14
UAP-HF-08019-14
QA PX
LOCA
V Statistical methodology of large break LOCA demonstrates
that acceptance criteria of 10 CFR 50.46 are satisfied
PCT(95/95) = 1763 OF < 2200 OF
V Conservative analysis of small break LOCA, which is based
on Appendix-K, demonstrates that acceptance criteria of
10 CFR 50.46 are satisfied
PCT = 1297 OF < 2200 OF
V Switchover to simultaneous RV and hot leg injection mode at
four hours after a LOCA prevents boric acid precipitation in
the core, then post-LOCA long term cooling is assured
IMMUIPSURISHI-HEA.Y7M-INU-STRES,
LTD.
UAP-HF-08019-I 5
UAP-HF-08019-15
6. Analysis Results (Cont'd)
Reactivity Initiated Accident (RIA) Specific Criteria
(SRP 4.2 Appendix B and SRP 15.4.8)
/ The average fuel pellet enthalpy at the hot spot remains
significantly below 230 cal/g
/ 3-D methodology is applied to analyze Rod Ejection Accident
from HZP condition. Prompt fuel enthalpy rise is well below
new threshold for cladding failure.
2O0
175
(0.04,150)
150
Cladding Failure
125
u. 100
0
75
(0.08, 75)
50
(0.20, 0)
0
0.04
0.
08
0.12
0.16
0.2
d./Wall Thick.*=
M-.uinumm 14AVY1NDUItRIES, LTD.
6. Analysis Results (Cont'd)
UAP-HF-08019-16
/-Us-
RADIOLOGICAL CONSEQUENCE ANALYSES
/ The exclusion area boundary (EAB) and the outer boundary
of low population zone (LPZ) doses are shown to meet the
10 CFR 50.34 dose guidelines
The dose results (LOCA) are 13rem < 25rem at EAB,
13rem < 25rem at LPZ
V The dose for the MCR personnel is shown to meet the dose
criteria given in GDC 19
The dose results (LOCA) are 4.5rem < 5rem in MCR
&.-MTr#tXB1%M HEAVY IMPUSTRIES, LTD.
L..MLTWW4$HI HEAVY INDUSTRIES, LTD.
UAP-HF-08019-17
9-17
UAP-HF-0801
7. Summary
AP
1. US-APWR DCD FSAR Chapter 15 format and
content comply with RG 1.206 and satisfy the SRP
requirements
2. All results of transient and accident analyses meet
the acceptance criteria
3. Methodologies and codes for US-APWR are
discussed in Topical Reports for NRC review
4. Supplemental information is provided in Technical
Report to support DCD review
L.
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UAP-HF-OBOIQ-18
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 16
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
; _MfI.SUBISHI-HEVJ-Y--INDUSlRIES, LTD.
UAP-HF-08020
Presenter
Katsunori Kawai
Engineering Manager
APWR Promoting Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, Ltd.
T.SýVW$""EW-Y-FIM
.TjkIIES, LTD.
~e~MIT5UBI5HU~HEAV-Y-INDU5jR!ES, LTD.
UAP-HF-08020-1
UAP-HF-08020-1
Contents
Contents
A ýPM
H
1. Overview of Chapter
Title of Chapter
Scope of Chapter
2. Features of Technical
Specifications (TS)
L~41$!BEL!EAVY4DUSRISLTD.
1.
Overview of Chapter
UAP-HF-08020-2
us -3_
'-ýW 10
Title of Chapter
Chapter 16: Technical Specifications
SScope of Chapter
*
*
This chapter includes the following categories
of information as required by 10 CFR 50.36 and
10 CFR 50.36a
Safety limits, limiting safety system settings,
LCOs, surveillance requirements, design
features and administrative controls
LMIIT.$!J-BISHI-Hgg-gC-Y_-4!-MWW$-T-IkIES, LTD.
~JNDUSTRIES, LTD.
MI~S.UBISHIHE
UAP-HF-08020-3
UAP-HF-08020-3
UAP-HF-08020-3
/-Uso/,-.-ý_ýt,,
AA
2. Features of TS
Features of US-APWR safety system design
0
Design concept is based on current
PWRs in the USA
*
Four-train safety systems are one of
characteristic design features
Features of US-APWR Technical
Specifications
* Basically follow the Standard TS* (STS)
* Maximize the benefits of on-line maintenance (OLM)
* Apply Risk-Managed Technical Specifications
*
NUREG-1431, Rev.03, "Standard Technical Specifications Westinghouse Plants"
LR!Th!AtIlkIU$EAV--IDU5JRIES,
LTD.
UAP-HF-08020-4
2. Features of TS (cont'd)
2.1 Utilization of STS
* US-APWR Technical Specifications are almost same
as the STS of NUREG-1431
* US-APWR Technical Specifications differ from STS
only as necessary to reflect technical differences
between conventional
US- PWRs design and US-APWR design
* Justification for deviations between STS and USAPWR TS is described in technical report *
•: Justification for Deviations between NUREG-1431 and
US--APWR Technical Specifications (Dec. 2007)
'- MEESUBISHI-HE-AV-Y•4•ND-U-S-TRIES, LTD.
UAP-HF-(08020-5
H
2. Features of TS (cont'd)
2.2 Safety Benefits of Four-train systems
* Enhanced redundancy (50% x 4)
vCapability beyond single failure criterion
* Maximize the benefits of on-line maintenance
v/Establish LCO requiring three trains
operable
v/Establish completion time when one of the
three required trains inoperable
MIISU BISHI-HEAV-Y-IND U STRIES, LTD.
UAP-HF-08020-6
2. Features of TS (cont'd)
W
2.3 Main deviations between STS (NUREG1431) and US-APWR TS
Characteristic design features
* Four train safety systems
- e.g. : LCO is Three of four SIS trains shall be
OPERABLE
* Gas turbine generators
- e.g. Fuel oil testing program
* Digital Platform
- e.g. Actuation logic test interval increased
Surveillance Interval
* 24 month refueling cycle
•M•I.T.SUBiSHIHEV=ND US "IES,
LTD.
UAP-HF-08020-7
- IF,
2. Features of TS (cont'd)
2.4 Adoption of Risk-Managed Technical
Specifications (RMTS)
Risk-Informed Completion Times (CTs)
* A front-stop CT and Commitment to
Configuration Risk Management Program
(CRMP)
* 30-day limit as a back-stop CT
*
Reference to Risk-Managed Technical
Specifications Initiative 4b*
•: NEI 06-09 (Revision 0) "Risk-Informed Technical Specifications
Initiative 4b Risk- Managed Technical Specifications (RMTS)
Guidelines," November 2006.
UAP-HF-08020-8
.MT.SUB
tISHi-HEkV.=HY-IU$SjR IES, LTD.
2. Features of TS (cont'd) .......
Coming works for RMTS to be completed
Establishment of the station procedure of
the Configuration Risk Management
Program (CRMP)
Training of responsible personnel
Preparation of a PRA model to meet the
technical adequacy requirement of NEI 0609
Preparation of an appropriate CRM tool
~MI~UBESI-HEV~I~MLSTIESLTD.
UAP-HF-08020-9
UPH-82UAP-HF-08020-9
2. Features of TS (cont'd)
QVPS5
Other risk-informed initiatives will be considered
Initiative 5b: Relocation of all SR frequency
requirements out of TS
> Initiative 1: Actions end states modification
$
This initiative would permit, for some system, entry into hot
shutdown rather than cold shut down to repair equipment
> Initiative 7: Non-TS support system impact on
TS operability determinations
V
This initiative would permit a risk-informed delay time before
entering LCO actions for Inoperability due to loss of support
function provided by equipment outside of technical
specifications
I;;ýMfl5rUBUH1_WEEAV-Y
UAP-HF-08020-1 0
UAP-HF-08020-10
IMDqSTRIES. LTD.
2. Features of TS (cont'd)
(SAP
.5111,
SstSubmitdat e~achstage in app~lying RMTS
P lao
Stage
DC
Tech. Spec.
(Incl. RMTS)
Design-specific DCD Chapter 19
Plant-specific
CL(abllshCd)s
Plant-specific PRA results consistent with
FSAR Chapter 19 to support RMTS
Plant-specific
(All CTs
established)
0 Technical report describing PRA
technical adequacy, CRM tools, CRMP,
Organization, Training of personnel, etc*
*Implementation manual
established)
Prior to
fuel load
Associated Documents
*All required ITAAC
*: In accordance with NEI 06-09
M--,jM9jQ$TR1ES, LTD.
*~MI.TSUBISIII-HEAV-Y-INDUSJRIES,
LTD.
UAP-HF-08020-11
UAP-HF-08020-1 I
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 17
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
L
-J_
DLISTRIES, LTD.
UAP-HF-08021
Presenter
Naoki Miyakoshi
General Manager
Nuclear Energy Systems Quality and Safety
Management Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, LTD.
&'=WT-SUB1SH11-HE-AM-, Y-UfflPV5jR1ES, LTD.
LTD.
UAP-HF-08021-1
UAP-HF-08021 -1
Contents
1. Overview of Chapter
Title of Chapter
v'
Scope of Chapter
2. Chapter 17 Contents
3. Summry
IIA[•
ýMITSUISHI
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HIE AVY
-ImNDUSTRIES
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1.
L TD
-LAr-r-1r-uouL
U•
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V-
,•D
1-4
Overview of Chapter
STitle of Chapter
Chapter 17: QUALITY ASSURANCE AND
RELIABILITY ASSURANCE
SScope of Chapter
Quality Assurance Program performed during
the design certification phase
Design Reliability Assurance Program (Phase I
D-RAP; design certification phase)
MITSUBISHI HEAVY INDUSTRIES, LTD.
UAP-HF-08021-3
1.
Overview of Chapter (cont'd)
usAzD-
Scope of Chapter (cont'd)
17.1
Quality Assurance During Certification Phase
DC Phase
17.2
Quality Assurance During the Construction and
(COL)
Operation Phase
•
17.3
Quality Assurance Program Description
DC Phase
17.4
Design Reliability Assurance Program
DC Phase
17.5
Quality Assurance Program Guidance
DC Phase
17.6
Description of the Applicant's Program for
Implementation of 10CFR 50.65, the
Maintenance Rule
(COL)
MITSUBISHI HEAVY INDUSTRIES, LTD.
UAP-HF-08021-4
2. Chapter 17 Conte nts
Quality Assurance Program
,/ QAP meets requirements of 1C]CFR Part 50,
Appendix B, 10CFR Part2I anc 10CFR Part52.
v/ QAP is based on the requirements of ASME
NQA-1 -1994 "Quality Assurance Requirements
for Nuclear Facilities Applications,"
Parts I and II.
v/ QAP Description for DC phase has been
prepared on the basis of the NRC approved
QAP template (NEI 06-14A Rev.4)
MITSUBISHI HEAVY INDUSTRIES, LTD.
UAP-HF-0)Rf021-5
8021-5
2. Chapter 17 Contents (cont'd)
> Quality Assurance Program
_
50vt
n App•ndix
' - 14n CF
-"--l
ato
IA Requirements
Applicable
Aertatutt,
/
/
1. Organization
2. QA Program
3. Design Control
4.
5.
6.
7.
_2
Remarks (MHI QAP on US-APWR)
/
Procurement Document Control
Instructions, Procedures and Drawings
Document Control
Control
of Purchased Materials, Items and
Rervinr_-•
v/
I/
-
8. Identification and Control of Items and
Materials
AtlchDC stage
this aplies
to services
_ n lv i
T=t
Not Applicable (NUREG-0800 17.5)
9. Control of Special Processes
10. Inspection
11. Test Control
12.
13.
14.
15.
16.
17.
At DC stage this applies to
inspections for test facilities
- At DC stage this applies to
qualification tests
-
,
Control of Measuring and Test Equipment
Handling, Storage and Shipping
Inspection, Test and Operating status
Control of Nonconforming Items
Corrective Action
QA Records
/: Comply
M18. Audit
MI~TSU BISHI HEAVY INDUSTRIES, LTD.
-: N/A
UAP-HF-08021-6
2. Chapter 17 Contents (cont'd)
WI
Reliability Assurance Program
v1 The sco e of DCD chapter 17 is Phase I D-RAP.
Phase I
Design Certification phase
Phase II
Site-specific phase
Phase III Last phase of D-RAP
(procurement, fabrication, construction
preoperational testing)
V US-APWR D-RAP identifies risk-significant
SSCs and provides risk insights and reliability
assumptions.
MITSUBISHI HEAVY INDUSTRIES, LTD.
UAP-HF-(08021-7
08021-7
2. Chapter 17 Contents (cont'd)
Q
Reliability Assurance Program (cont'd)
SResponsibility (Phase I D-RAP)
v' General Manager, APWR project:
- Establishment of US-APWR D-RAP program
-/General Manager, Reactor and Plant Safety:
- Use of the PRA results and risk insights for the
Expert Panel
- Conduct and coordination of the Expert Panel
V General Manager, QA:
- Assuring proper implementation of QA program
(Organization, design control, procedure and
instruction, records, corrective actions, audit)
MITSUBISHI HEAVY INDUSTRIES, LTD.
UAP-HF-08021-8
2. Chapter 17 Contents (cont'd)
List of risk-significant SSCs
/ The risk and reliability organization is
responsible to provide the RAP related inputs in
the design process.
V List of risk-significant SSCs is initially based on
the result of PRA and Expert Panel.
$ The list and changes shall be approved by
Expert Panel.
/ List of risk-significant SSCs and its key
assumptions shall be maintained by the risk
and reliability organization.
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3. Summary
Q
MHI established QAPD for Design Certification
and carried out design activities in accordance
with the QAPD. MHI submitted the QAPD as a
topical report to contribute for the NRC review.
Topical Report
"Quality Assurance Program (QAP) Description
For Design Certification of the US-APWR (PQD-HD19005 Rev.1)"
MHI established D-RAP (phase 1)program and
prepared a list of risk-significant SSCs.
MITSUBISHI HEAVY INDUSTRIES, LTD.
UAP-HF-08021-10
US-APWR
Design Certification Application Orientation
Detail of FSAR
Tier2: Chapter 19'
January 15,16, 2008
Mitsubishi Heavy Industries, Ltd.
IS•UBISHi=HE~V-•Y
U-s•T•RIES, LTD.
UAP-HF-08022
Presenter
Katsuya Kuroiwa
Engineering Manager
Reactor Safety Engineering Department
Nuclear Energy Systems Headquarters
Mitsubishi Heavy Industries, Ltd.
ýD-U-$Rk1E S, LTD.
MlURTSUBIS-HI-HEAV. -Y- INW
UAP-HF-08022-1
UPH-82-
Contents
QA A
1. Overview of Chapter
> Title of Chapter
> Scope of Chapter
2. Probabilistic Risk Assessment
3. Severe Accident Evaluation
4. Summary
K MIEMEDnCUS
U=Alff% 1WflhE1EC'rn1WQ
1.
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I IAP-HI=-lNfn79-9
UAP-HF-08022-2
/_VS7ý1;h
Overview of Chapter
Title of Chapter
Chapter 19: PROBABILISTIC RISK ASSESSMENT
AND SEVERE ACCIDENT EVALUATION
SScope of Chapter
Probabilistic Risk Assessment (PRA) results
and insights including internal and external
events during full-power operations and during
low power and shutdown operations
Severe accident evaluations including an
assessment of preventive and mitigation
features, containment performance capability,
accident management and severe accident
mitigation design alternatives (SAMDA)
_Yit qW!kr
RýJES, LTD.
_=V
~MEISUBISHIHEAVYINDUSTRIES. LTD.
UAP-HF-08022-3
UAP-HF-08022-3
2. Probabilistic Risk Assessment
CA P-S;b,
> Methods
and Approach
.Basic design concept
of US-APWR is similar to current
PWRs. Therefore, present guides and standards are applied.
,/Regulatory Guide 1.200 Rev.1
,/Standards endorsed by Regulatory Guide 1.200
Rev.1
* ASME RA-S-2002 and the addenda ASME
RA-Sa-2003, ASME RA-Sb-2005
* ANSI-ANS 58.21-2003
,/Areas where no formal standards exists, previous
studies or guidance are used
MiI--UIjSHIHEA-V.Y-INDUPSIRIES,
LTD.
UAP-HF-08022-4
2. Probabilistic Risk Assessment (cont'd)
A
-Special Design Features of US-APWR
Improved plant safety as compared to currently
operating nuclear power plants
*
Higher redundancy: four train mechanical and
electrical safety systems
• Simplicity: In-containment RWSP eliminates
recirculation switchover
* Independent: Physical separation of four train
safety systems
* Diversity: Alternative systems such as diverse
actuation system, alternative AC power source
etc.
MIXISURISM-HkAV=Y=1
RkU-S-TRIES, LTD.
LTD.
UAP-HF-08022-5
UAP-HF-08022-5
2. Probabilistic Risk Assessment (cont'd)
> Level 1 Internal Events PRA at Power
* Core Damage Frequency (CDF): 1.2 x 10-61RY
FWLS
LOAC
0'4% -02
ATWS
1.2%
SGT
06%
LOOC
L
01%
I
TRANS
Large Pipe Break LOCA
MLOCA
Medium Pipe Break LOCA
SLOCA
Small Pipe Break LOCA
VSLOCA
Very Small Pipe Break LOCA
01%
LLOCA
SGTR
Steam Generator Tube Rupture
0 04%
RVR
VSLOCA
12%
PLOCW
13%
LLOCA
FWLB
Reactor Vessel Rupture
Steam Line Break/Leak
(Downstream MSIV: Turbine side)
Steam Line Break/Leak
(Upstream MSIV: CV side)
Feed-water Line Break
TRANS
General Transient
LOFF
Loss of Feed-water Flow
LOCCW
Loss of Component Cooling Water
PLOCW
Partial Loss of Component Cooling Water
LOOP
Loss of Offsite Power
LOCCW
LOAC
Loss of Vital ac Bus
25 6%
LODC
Loss of Vital DC Bus
MLOCA
1 4%
1 6%
LOFF
16%
RVR
49.3%
65%
Core Damage Frequency Contribution
HEAVY INDUSTRIES, LTD.
"T-SUDaISH
UAP-HF-08022-6
2. Probabilistic Risk Assessment (cont'd)
Q--W
Level 2 Internal Events PRA at Power
* Large Release Frequency (LRF) : 1.0 x 10- 7IRY
SLBO,
09%
RVR
TRANS VSLOCA FWLB AIWS
0 8%
0 6%
02%01%
LOAC
01%
5L51
SL%
LOC
11%
LOFF
001%-
13%
LLOCA
00%
MLOCA
2 1%
SGTR
Large Pipe Break
MLOCA
SLOCA
Medium Pipe Break LOCA
Small Pipe Break LOCA
VSLOCA
Very Small Pipe Break LOCA
.GTR
SGTR
.Generator.
Tube.Rup.. .
.Steam
Steam Generator Tube Rupture
RVR
SLBO
Reactor Vessel Rupture
Steam Line Break/Leak
(Downstream MSIV: Turbine side)
SLBI
(Upstream
MSIV:
-am=. ne=
rea CV
ea=side)
FWLB
Feed-water Line Break
TRANS
General Transient
LOFF
Loss of Feed-water Flow
LOCCW
Loss of Component Cooling Water
PLOCW
Partial Loss of Component Cooling Water
LOOP
Loss of Otfsite Power
LOAC
Loss of Vital ac Bus
LODC
Loss of Vital DC Bus
60%
PLLOC
346%
29,4%
LOCA
LLOCA
1,
.
-
Large Release Frequency Contribution
A-*w
P41TSUBISHI HEAVY INPUSTRIES, LTD.
.~,MLT#IJmNJ 4IM~Y INDUSTRIES. LTD.
UAP-HF-08022-7
UAP-H F-08022-7
2. Probabilistic Risk Assessment (cont'd)
Uncertainty Analysis for Internal Events PRA
at Power
1.0E-05
1.0E-O6
:
95percenfile;
3.OE-07
2-9E-06 95percentile
0
1.2E-06
Mean
7.8E-07
Median
-77------ ------Mean; 11 E-07
1.OE-07
1.OE-06
Median; 6.5E-08
3,OE-07
5percentile
5 percentile;
2.DE-08
1.0E-07
Core Damage Frequency
Large Release Frequency
UAP-HF-08022-8
•AjIPIOl'NSill HEAVY INPUSTRIES, LTD.
2. Probabilistic Risk Assessment (cont'd)
o PRA Results of Other Events
CDF
Seismic
LRF
(Seismic Margin Analysis)
Plant HCLPF: .0.5g
Internal Fire
1.7xl0-6/RY
2.0xl0 7/RY
Internal Flood
1.5x1 0 6/RY
4.0xl 0 7/RY
Other External
Low Power and
Shutdown
Site Specific
2.0xl07/Ry
00HUHIFAVY INDUSTRIES, LTD.
lRI$NI HEAVY INDUSTRIES LTD.
Assumed to be
same with CDF
UAP-HF-08022-9
UAP-HF-08022-9
2. Probabilistic Risk Assessment (cont'd)
> Risk Significant Scenario and SSCs
* Station blackout with common cause failure of
emergency gas turbine generators
+ Alternative AC power is effective to reduce the risk
" Loss of component cooling water by common cause
failure of component cooling water pumps
+ Independent trains are effective to eliminate the risk of
loss of cooling water by leak
* Fire in the switchyard area, causes loss of offsite power
+ Physical separation is effective to reduce the risk of fire in
other areas
* Major flood in the divided area of the reactor building,
causes partial loss of safety functions
+ Physical separation is effective to reduce the risk
_M
I.S U B I
,H-EAV-Y-I
ND-U-S-TR IES, LTD.
UAP-HF-08022-10
2. Probabilistic Risk Assessment (cont'd)
> PRA Insights and Design Features
* CDF and LRF are less than the NRC goals
( less than I E-4/year for CDF and less than
1 E-6/year for LRF)
* Design features of US-APWR as shown below
reduce the risk.
+ Four train safety systems
+ Independent four train electrical system with
alternative AC power source
Iin-containment RWSP
+ Various severe accident prevention/mitigation
features
Z-ý,MI.X.SýURISH-1-HE-AX-Y=IMCtUS-FgklES,
LTD.
LTD.
UAP-HF-08022-11I
UAP-HF-08022-1
2. Probabilistic Risk Assessment (cont'd)
Risk-informed Applications at Design Phase
*
PRA has been used to optimize the plant
design with respect to safety.
• Assumptions of important operating actions
are identified for the accident management
framework.
* Risk significant SSCs are identified for the
Reliability Assurance Program (RAP).
*
PRA insights are utilized to develop riskmanaged technical specifications (RMTS).
MflTSBISHI-HEAVY=INMWUSTRIES, LTD.
UAP-HF-08022-12
3. Severe Accident Evaluation
Prevention and Mitigation
• Apply proven techniques for existing plants with
improvements
Analysis Approaches and Methods
* Apply analysis approaches accepted by NRC for
former DC applications
" Employ MAAP4.0.6 for severe accident
progression analysis, and other specific codes
for specific phenomena
- jWkt
LCALTiSUBIj-IHKAV~INPJR
ES, LTD.
UAP-HF-08022-13
UPH-82-1
UAP-HF-08022-13
3. Severe Accident Evaluation (cont'd)
1-UPS5-!#A
Severe Accident Prevention Features
Anticipated Transient - Four train reactor protection system
- Diverse actuation system
Without Scram
- Automatic let-down isolation
Mid-Loop Operation
- Alternative core cooling
- Four emergency gas turbine generators
Station Blackout
- Two alternative AC power sources
- Physically separated four train safety
Fire Protection
systems
Intersystem Loss-ofCoolant Accident
Others
~i.MUW
- Up-rated RHRS piping
- Feed and bleed with redundancy
- Alternative component cooling, etc.
H HEAVY INDU*TRIES, LTD.
UAP-HF-08022-14
3. Severe Accident Evaluation (cont'd)
Severe Accident Mitigation Features
4-7JS5-P
_4
Addressed
severe accident
issues
(1) Hydrogen
generation and
control
(2) Core debris
coolability
(3) Steam
explosion
(4) HPME
(5) TISGTR
(6) MCCI
(7) Long-term
containment
overpressure
(8) Equipment
survivability
HEAVY INDUSTRIES, LTD.UAHF00-5
Am AWR# 5111
UAP-HF-08022-15
3. Severe Accident Evaluation (cont'd)
(!APSIR
- Containment Performance (SECY-93-087)
*
Deterministic goal:
÷ Containment integrity be maintained for
approximately 24 hours following the onset
of core damage for the more likely severe
accident challenges
* Results:
÷ Containment integrity is maintained for
more than 24 hours following the onset of
core damage for most of the severe
accident conditions
-.MI.TUBISHIHEAV-Y-INDUS TRIES. LTD.
UAP-HF-08022-16
3. Severe Accident Evaluation (cont'd)
(P
I Accident Management
* Develop a framework includes:
÷ Approach
÷ Operational and phenomenological conditions
÷ Basis of the actions
* Four countermeasures and operating actions:
+ To prevent core damage
+ To terminate the progress of core damage if it
begins and to retain the core within the reactor,
vessel
÷ To maintain containment integrity as long as
possible
÷ To minimize offsite release
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3. Severe Accident Evaluation (c ont'd)
AP
> SAMDA
*
Meet the requirement of 10 CFR 50.34(f)(1)(1) to
consider potential design improvements
Approaches:
-
Guidance for regulatory analysis (NUREG/BR-0184
NUREG/BR-0058)
+ Industry implementation guidance (NEI 05-01, Rev. A)
- consistent with SECY-99-169
*
and
Results:
+ Ten candidate SAMDAs are selected from 156 potential
improvements
+ The benefit of each SAMDA is observed to be significantly
less than the cost impact
*
No additional design alternatives are shown to be costbeneficial in severe accident mitigation design
LM:TSUBIS5H1,aEAV-Y-INDUSTRIES,
LTD.
UAP-HF-08022-18
4. Summary
(-ALPW34-
" Describe the design-specific PRA
* PRA results indicate the US-APWR design meets
the NRC safety goals.
* Describe design features for the
prevention and mitigation of severe
accidents
!-MI.,SUBI
HIEAVY-INDUSTRIES
I LTD.
UAP-FIF-08022-19